ML25071A389

Issuance of Amendment Nos. 365 and 346 Regarding Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

ML25071A389
Person / Time
Site:	Cook
Issue date:	04/09/2025
From:	Scott Wall Plant Licensing Branch III
To:	Ferneau K Indiana Michigan Power Co
Wall S
References
EPID L-2023-LLA-0080
Download: ML25071A389 (41)

Text

April 9, 2025 Kelly J. Ferneau Executive Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 365 AND 346 REGARDING ADOPTION OF 10 CFR 50.69, RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS AND COMPONENTS FOR NUCLEAR POWER REACTORS (EPID L-2024-LLA-0025)

Dear Kelly J. Ferneau:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 365 and 346 to Renewed Facility Operating License (RFOL) Nos. DPR-58 and DPR-74, for the Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP), respectively. The amendments consist of changes to the RFOLs in response to your application dated March 6, 2024, as supplemented by letter dated February 6, 2025.

The amendments revise the CNP RFOL Nos. DPR-58 and DPR-74 to add a new license condition to allow for implementation of Title 10 of the Code of Federal Regulations Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Scott P. Wall, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosures:

1. Amendment No. 365 to DPR-58

2. Amendment No. 346 to DPR-74

3. Safety Evaluation cc: Listserv

INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 365 License No. DPR-58

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana Michigan Power Company dated March 6, 2024, as supplemented by letter dated February 6, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. DPR-58 is hereby amended to add paragraph 2.C.(21) to read as follows:

(21) 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)

RISC-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. 365 dated April 9, 2025.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: April 9, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.04.09 15:16:49 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 365 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-315 Renewed Facility Operating License No. DPR-58 Replace the following page of the Renewed Facility Operating License No. DPR-58 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

REMOVE INSERT Renewed License No. DPR-58 Amendment No: 315, 319, 322, 323, 325, 333, 334, 365 (19) Operation with Vacuum Fill:

The licensee is authorized to operate the facility using Reactor Coolant System (RCS) vacuum fill operation in accordance with TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, with corresponding revisions to Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY), and Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY), as approved in License Amendment No. 323 to Renewed Facility Operating License No. DPR-58. This includes an approved extension to -14.7 pounds per square inch gage to bound the RCS conditions required to support vacuum fill operation. The licensee shall submit an analysis of the P/T curves in Figures 3.4.3-1 and 3.4.3-2 within one year of the date of issuance of License Amendment No. 323, which demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G to 10 CFR Part 50, including non-beltline ferritic reactor vessel materials.

(20) The licensee shall implement the items listed in Enclosure 2, Table 1, of I&M letter AEP-NRC-2016-69, dated September 9, 2016, prior to Surveillance Frequency Control Program implementation.

(21) 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. 365 dated April 9, 2025.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

D. Physical Protection The Indiana Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revision to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and Renewed License No. DPR-58 Amendment No: 319, 322, 323, 334, 365 10 CFR 50.54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Donald C. Cook Nuclear Plant Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 1, submitted by letter dated May 10, 2006.

The Indiana and Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved Donald C. Cook Nuclear Plant Cyber Security Plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Donald C. Cook Nuclear Plant CSP was approved by License Amendment No. 315 as supplemented by changes approved by License Amendment Nos. 319, 325, and 333.

E.

Deleted by Amendment No. 80 F.

Deleted by Amendment No. 80 G.

In all places of this renewed operating license, the reference to the Indiana and Michigan Electric Company is amended to read Indiana Michigan Power Company.

H.

Deleted by Amendment No. 287 I.

Deleted by Amendment No. 287 J.

The licensee is authorized to use digital signal processing instrumentation in the reactor protection system.

K.

Updated Final. Safety Analysis Report The Indiana Michigan Power Company Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. The Indiana Michigan Power Company shall complete these activities no later than October 25, 2014, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed operating license. Until that update is complete, Indiana Michigan Power Company may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Indiana Michigan Power Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

J.

All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan Renewed License No. DPR-58 Amendment No: 319, 322, 323, 334, 365 withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

3. This renewed operating license is effective as of the date of issuance and shall expire at midnight, October 25, 2034.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Appendix A - Technical Specifications

2.

Appendix B - Environmental Technical Specifications Date of Issuance: August 30, 2005

INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 346 License No. DPR-74

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana Michigan Power Company dated March 6, 2024, as supplemented by letter dated February 6, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. DPR-58 is hereby amended to add paragraph 2.C.(3)(ii) to read as follows:

(ii) 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)

RISC-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. 346 dated April 9, 2025.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: April 9, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.04.09 15:17:16 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 346 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-316 Renewed Facility Operating License No. DPR-74 Replace the following page of the Renewed Facility Operating License No. DPR-74 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

REMOVE INSERT Renewed License No. DPR-74 Amendment No. 299, 303, 305, 306, 308, 315, 316, 346 (III)

The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test, or within 182 days if not performed previously.

(gg) Operation with Vacuum Fill:

The licensee is authorized to operate the facility using Reactor Coolant System (RCS) vacuum fill operation in accordance with TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, with corresponding revisions to Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits

- Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY), and Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY), as approved in License Amendment No. 306 to Renewed Facility Operating License No. DPR-74.

This includes an approved extension to -14.7 pounds per square inch gage to bound the RCS conditions required to support vacuum fill operation. The licensee shall submit an analysis of the P/T curves in Figures 3.4.3-1 and 3.4.3-2 within one year of the date of issuance of License Amendment No.

306, which demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G to 10 CFR Part 50, including non-beltline ferritic reactor vessel materials.

(hh) The licensee shall implement the items listed in Enclosure 2, Table 1, of I&M letter AEP-NRC-2016-69, dated September 9, 2016, prior to Surveillance Frequency Control Program implementation.

(ii) 10 CFR 50.69 Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. 346 dated April 9, 2025.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Renewed License No. DPR-74 Amendment No. 305, 306, 316, 346 D. Physical Protection The Indiana Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revision to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Donald C. Cook Nuclear Plant Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 1, submitted by letter dated May 10, 2006.

The Indiana and Michigan Power Company shall fully implement and maintain in effect all provisions of the Commission-approved Donald C. Cook Nuclear Plant Cyber Security Plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Donald C. Cook Nuclear Plant CSP was approved by License Amendment No. 315 as supplemented by changes approved by License Amendment Nos. 303, 308, and 315.

E.

Deleted by Amendment No. 63 F.

In all places of this renewed operating license, the reference to the Indiana and Michigan Electric Company is amended to read Indiana Michigan Power Company.

G.

Deleted by Amendment No. 269 H.

Deleted by Amendment No. 269 I.

Deleted by Amendment No. 261 (1) Deleted by Amendment No. 261 (1) Deleted by Amendment No. 261 J.

The licensee is authorized to use digital signal processing instrumentation in the reactor protection system.

K.

Updated Final. Safety Analysis Report The Indiana Michigan Power Company Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21(d), describes certain future activities to be completed prior to the period of extended operation. The Indiana Michigan Power Company shall complete these activities no later than December 23, 2017, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4) following issuance of this renewed operating license. Until that 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan Renewed License No. DPR-74 Amendment No. 305, 306, 316, 346 update is complete, Indiana Michigan Power Company may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Indiana Michigan Power Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

L.

All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion.

3. This renewed operating license is effective as of the date of issuance and shall expire at midnight, December 23, 2037.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1.

Preoperational Tests, Startup Tests and Other Items Which Must Be Completed Prior to Proceeding to Succeeding Operational Modes.

2.

Appendix A - Technical Specifications

3.

Appendix B - Environmental Technical Specifications Date of Issuance: August 30, 2005

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 365 AND 346 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-58 AND DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316

1.0 INTRODUCTION

By letter dated March 6, 2024 (Reference 1), as supplemented by letter dated February 6, 2025 (Reference 2), Indiana Michigan Power Company (I&M, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP). The amendments would allow the licensee to implement Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.

The provisions of 10 CFR 50.69 allow adjustment of the scope of structures, systems, and components (SSCs) subject to special treatment requirements (e.g., quality assurance, testing, inspection, condition monitoring, assessment, and evaluation). The adjustment is based on an integrated, systematic, risk-informed process that includes several approaches and methods for categorizing SSCs according to their safety significance.

The licensee proposed the following license condition for CNP Renewed Facility Operating License (RFOL) Nos. DPR-58 and DPR-74:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS [American Society of Mechanical Engineers/American Nuclear Society] PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. [XXX] dated April 9, 2025.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The NRC staff participated in a regulatory audit from May 13, 2024 to January 3, 2025 to ascertain the information needed to support its review of the application and develop requests for additional information, as needed. In email correspondence dated December 30, 2024, the NRC staff requested additional information from the licensee (Reference 3). The licensee responded to the request for additional information (RAI) in a supplemental letter dated February 6, 2025 (Reference 2).

The supplemental letter dated February 6, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 14, 2024 (89 FR 41999).

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations The provisions of 10 CFR 50.69 allow adjustment of the scope of SSCs subject to special treatment requirements. Special treatment refers to those requirements that provide increased assurance (beyond normal industry practices) that SSCs will perform their design-basis functions. For SSCs with a function that is categorized as low safety significant (LSS),

alternative treatment requirements may be implemented in accordance with the regulation. For SSCs with a function determined to be high safety significant (HSS), requirements may not be changed.

Section 50.69 of 10 CFR contains requirements regarding how a licensee categorizes SSCs using a risk-informed process; adjusts treatment requirements consistent with the relative significance of the SSC; and manages the process over the lifetime of the plant. A risk-informed process is employed to determine the safety significance of SSCs and assign each into one of four RISC categories.

SSC categorization does not allow for the elimination of SSC functional requirements or allow equipment that is required by the deterministic design basis to be removed from the facility.

Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as HSS, existing treatment requirements are maintained and may be enhanced. Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows an alternative, risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on HSS equipment.

2.2 Regulatory Guidance The NRC staff considered the following regulatory guidance during its review of the proposed changes:

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 4)

RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 5)

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 6)

RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 7)

RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (Reference 8)

NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking (Reference 9)

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP) Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance (Reference 10)

Federal Register notice (69 FRN 68008, 68028-68029), dated November 22, 2004, related to Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors (Reference 30) 2.3 NRC-Endorsed Guidance The Nuclear Energy Institute (NEI) issued NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (Reference 11), as endorsed by RG 1.201, Revision 1 (Reference 8),

for trial use with clarifications.3 The guideline describes a process that the NRC staff considers acceptable for complying with 10 CFR 50.69. This process uses PRA and other methods to determine the safety significance of SSCs and categorizes them into one of four RISC categories defined in 10 CFR 50.69.

3 Regulatory Guide 1.201 (Reference 8) describes SSC categorization as an integrated decision-making process that incorporates both PRA and non-PRA evaluations of safety significance. Licensees must use risk evaluations and insights that cover the full spectrum of potential events and the entire range of plant operating modes.

Sections 2 through 10 of NEI 00-04 (Reference 11) describe the following steps/elements of the SSC categorization process for meeting the requirements of 10 CFR 50.69:

Sections 3.2, Use of Risk Information, and 5.1, Internal Events Assessment, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(i), which requires that the process consider results and insights from a plant-specific PRA.

Sections 3, Assembly of Plant-Specific Inputs; 4, System Engineering Assessment; 5, Component Safety Significant Assessment; and 7, Preliminary Engineering Categorization of Functions, provide specific guidance corresponding to 10 CFR 50.69(c)(1)(ii), which requires an integrated, systematic process.

Section 6, Defense-in-Depth Assessment, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iii), which requires maintaining defense-in-depth.

Section 8, Risk Sensitivity Study, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(iv), which requires maintaining sufficient safety margins and ensuring increases in risk are small.

Section 2, Overview of Categorization Process, provides specific guidance corresponding to 10 CFR 50.69(c)(1)(v), which requires that the process address entire systems and structures rather than selected components.

Sections 9, IDP [Integrated Decision-Making Panel] Review and Approval; and 10, SSC Categorization, provide specific guidance corresponding to 10 CFR 50.69(c)(2).

Section 11, Program Documentation and Change Control, provides guidance on program documentation and change control related to the requirements of 10 CFR 50.69(f).

Section 12, Periodic Review, provides guidance on the periodic review related to the requirements in 10 CFR 50.69(e), Feedback and process adjustment.

Maintaining change control and conducting periodic reviews provides confidence that all aspects of the program reasonably reflect both the current as-built, as-operated plant configuration and applicable plant and industry operational experience as required by 10 CFR 50.69(c)(1)(ii).

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review An acceptable approach for making risk-informed decisions about proposed changes, including both permanent and temporary changes, is to show that the proposed changes to the licensing basis meet the five key principles of risk-informed decision-making stated in Section C of RG 1.174, Revision 3 (Reference 4). These key principles are:

Principle 1:

The proposed licensing basis change meets the current regulations unless it is explicitly related to a requested exemption.

Principle 2:

The proposed licensing basis change is consistent with the defense-in-depth philosophy.

Principle 3:

The proposed licensing basis change maintains sufficient safety margins.

Principle 4:

When proposed licensing basis changes result in an increase in risk, the increases should be small and consistent with the intent of the Commissions policy statement on safety goals for the operations of nuclear power plants.

Principle 5:

The impact of the proposed licensing basis change should be monitored using performance measurement strategies.

The NRC staff evaluates the first three principles using traditional engineering methods. The last two are evaluated considering the licensees PRA and other methods of assessing risk.

3.2 Overview of Categorization Process Paragraph 50.69(c) of 10 CFR requires licensees to use an integrated decision-making process to categorize safety-related and non-safety-related SSCs according to the safety significance of the functions they perform. They are placed into one of the following four RISC categories:

RISC-1: Safety-related SSCs that perform safety significant functions4 RISC-2: Non-safety-related SSCs that perform safety significant functions RISC-3: Safety-related SSCs that perform low safety significant functions RISC-4: Non-safety-related SSCs that perform low safety significant functions The SSCs have functions that are HSS or LSS, and they are classified accordingly. For SSCs with HSS functions (i.e., RISC-1 and RISC-2 SSCs), 10 CFR 50.69 maintains current regulatory requirements for special treatment, that is, all existing special treatment requirements remain in force. In addition, 10 CFR 50.69(d)(1) requires that the licensee or applicant shall ensure that RISC-1 and RISC-2 SSCs perform their functions consistent with the categorization process assumptions by evaluating treatment being applied to these SSCs to ensure that it supports the key assumptions in the categorization process that relate to their assumed performance. For SSCs with LSS functions, licensees may implement alternative treatment requirements in accordance with 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2). For RISC-3 SSCs, licensees may replace special treatment requirements with an alternative treatment approach that meets 10 CFR 50.69(d)(2). For RISC4 SSCs, 10 CFR 50.69 does not impose new treatment requirements.

Section 50.69(b)(3) of 10 CFR states that the Commission will approve a licensees implementation of this section by issuance of a license amendment if the Commission 4 NEI 00-04 10, Revision 0, uses the term high-safety-significant to refer to SSCs that perform safety-significant functions. The NRC understands HSS to have the same meaning as safety-significant as used in 10 CFR 50.69, which applies to RISC-1 and RISC-2 SSCs.

determines that the categorization process satisfies the requirements of 10 CFR 50.69(c). As stated in 10 CFR 50.69(b), after the NRC approves an application for a license amendment, a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for LSS SSCs:

(i) 10 CFR Part 21 (ii) the specified portion of 10 CFR 50.46a(b)

(iii) 10 CFR 50.49 (iv) 10 CFR 50.55(e)

(v) specified requirements of 10 CFR 50.55a (vi) 10 CFR 50.65, except for paragraph (a)(4)

(vii) 10 CFR 50.72 (viii) 10 CFR 50.73 (ix)

Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 (x) specified requirements for containment leakage testing (xi) specified requirements of Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100 The NRC staff reviewed the licensees SSC categorization process against the categorization process described in NEI 00-04 (Reference 11), and the acceptability of the licensees PRA for use in the application of the 10 CFR 50.69 categorization process. The NRC staffs review, as documented in this safety evaluation, used the framework provided in RG 1.174, Revision 3 (Reference 4), and RG 1.201, Revision 1 (Reference 8).

Section 2 of NEI 00-04 states that the categorization process includes the following eight primary steps:

1.

Assembly of Plant-Specific Inputs

2.

System Engineering Assessment

3.

Component Safety Significance Assessment

4.

Defense-in-Depth Assessment

5.

Preliminary Engineering Categorization of Functions

6.

Risk Sensitivity Study

7.

Integrated Decision-Making Panel Review and Approval

8.

SSC Categorization 3.3 Traditional Engineering Evaluation The traditional engineering evaluation below addresses the first three key principles identified in RG 1.174, Revision 3 (Reference 4) and are pertinent to: (1) compliance with current regulations, (2) evaluation of defense-in-depth, and (3) evaluation of safety margins.

3.3.1 Key Principle 1: Licensing Bases Change Meets the Current Regulations In Section 3.1.1, Overall Categorization Process, of Enclosure 2 to the LAR (Reference 1), the licensee states that it will implement the risk categorization process in accordance with NEI 00-04, Revision 0 (Reference 11), as endorsed by RG 1.201, Revision 1 (Reference 8). The licensee provided further discussion of specific elements within the 10 CFR 50.69 categorization process that are delineated in the endorsed guidance of NEI 00-04.

The regulatory requirements in 10 CFR 50.69 and 10 CFR Part 50, Appendix B, as well as the monitoring outlined in NEI 00-04 (Reference 11), will ensure that the SSC functions continue to be met, that any performance deficiencies will be identified, and that appropriate corrective actions will be taken. The NRC staff finds that the licensees SSC categorization program includes the appropriate steps/elements prescribed in NEI 00-04 to assure that SSCs are appropriately categorized, consistent with 10 CFR 50.69. Therefore, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the first key principle for risk-informed decision-making identified in RG 1.174, Revision 3.

3.3.2 Key Principle 2: Licensing Basis Change is Consistent with the Defense-In-Depth Philosophy In RG 1.174, Revision 3, the NRC staff identified the following considerations used for evaluating whether a proposed licensing basis change is consistent with the defense-in-depth philosophy:

Preserve a reasonable balance among the layers of defense.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

Preserve adequate defense against potential CCFs [common-cause failures].

Maintain multiple fission product barriers.

Preserve sufficient defense against human errors.

Continue to meet the intent of the plants design criteria.

In Section 3.1.1 of the LAR enclosure (Reference 1), the licensee clarified that, consistent with the guidance in NEI 00-04 (Reference 11), it would require an SSC categorized as HSS based on the defense-in-depth assessment to be categorized HSS per the final categorization, and that cannot be changed by the IDP.

The NRC staff finds that the licensees process is consistent with the NRC-endorsed guidance in NEI 00-04 and concludes that the proposed change is consistent with the defense-in-depth philosophy. For this reason, the NRC staff concludes that the proposed 10 CFR 50.69 program meets the second key principle for risk-informed decision-making identified in RG 1.174, Revision 3 and fulfills the requirement of 10 CFR 50.69(c)(1)(iii) that defense-in-depth be maintained.

3.3.3 Key Principle 3: Licensing Basis Change Maintains Sufficient Safety Margins The third key risk-informed principle in RG 1.174 states that the licensing basis change should maintain sufficient safety margins. The engineering evaluation that will be conducted by the licensee under 10 CFR 50.69 for SSC categorization will assess the design function(s) and risk significance of SSCs to assure that sufficient safety margins are maintained. The guidelines used for making that assessment will include ensuring the categorization of the SSC does not adversely affect any assumptions or inputs to the safety analysis; or, if such inputs are affected, providing justification that sufficient safety margin will continue to exist.

The NRC staff notes that the design-basis functions of SSCs as described in the plants licensing basis, including the CNP Updated Final Safety Analysis Report and plant technical specification (TS) bases, do not change and the safety margins described should continue to be met. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. On this basis, the NRC staff concludes that safety margins are maintained by the proposed methodology, and the third key safety principle identified in RG 1.174, Revision 3 (Reference 4) is satisfied.

In Section 2.2, Reason for the Proposed Change, of the LAR enclosure (Reference 1), the licensee states, The safety functions [in the categorization process] include the design basis functions, as well as functions credited for severe accidents (including external events).

Section 3.1.1 of the LAR enclosure summarizes the different hazards and plant states for which functional and risk significant information will be collected. In the same section, the licensee confirmed that the SSC categorization process documentation will include, among other items, system functions, identified and categorized with the associated bases, and mapping of components to supported function(s).

The NRC staff finds that the process described in the LAR is consistent with NEI 00-04 as endorsed by the NRC in RG 1.201, Revision 1 (Reference 8). Therefore, it meets the requirements set forth in 10 CFR 50.69(c)(1)(ii) and 10 CFR 50.69(c)(1)(iv).

3.4 Risk-Informed Evaluation 3.4.1 Key Principle 4: Change in Risk is Consistent with the Safety Goals The risk-informed considerations prescribed in NEI 00-04 (Reference 11) address the fourth key principles of risk-informed decision-making identified in RG 1.174, Revision 3 (Reference 4). A summary of how the licensees SSC categorization process is consistent with the guidance and methodology in NEI 00-04, Revision 0 (Reference 11), and RG 1.201, Revision 1 (Reference 8),

is provided in the sections below. The NRC staff acknowledges that elements of the categorization process are not always performed in chronological order and may be performed in parallel.

3.4.2 Probabilistic Risk Assessment 3.4.2.1 Scope of the PRA In Section 3.2, Technical Adequacy Evaluation (10 CFR 50.69(b)(2)(ii)), of Enclosure 2 to the LAR (Reference 1), the licensee described the CNP PRA models, which include: a full-power internal events (FPIE) PRA (including internal floods), a fire PRA, and a seismic PRA. Each PRA evaluates risk metrics of core damage frequency (CDF) and large early release frequency (LERF). The licensee also states that the PRA models have been independently peer reviewed and there are no PRA upgrades that have not been peer reviewed.

In Section 3.3, PRA Review Process Results (10 CFR 50.69(b)(2)(iii)), the licensee provided a description of the peer reviews and the associated peer review Facts and Observations (F&O) findings closure reviews performed for the PRA models. F&O closure reviews were performed using the NRC-accepted process documented in the NEI letter to the NRC, Final Revision of Appendix X to NEI 0504/0712/1216, Close-out of Facts and Observations (F&Os)

(Appendix X) dated February 21, 2017 (Reference 13). In Attachment 3, Disposition and resolution of open peer review findings and self-assessment open items, of Enclosure 2 to the LAR, the licensee described the open F&O findings and provided a disposition for each of them.

The information in the LAR related to the peer reviews and disposition of F&Os was supplemented by the information in the licensees RAI responses (Reference 2). The NRC staffs evaluations of those peer reviews, closure reviews, and open findings are discussed in Section 3.4.2.2, below.

Aspects considered by the NRC staff to evaluate the scope of the PRA include: (1) a process for peer review and independent assessment, (2) history of peer reviews and their results, (3) credit taken in the PRA for the diverse and flexible coping strategy (FLEX), and (4) assessment of assumptions and approximations.

The information provided in the LAR (Reference 1) and the supplemental letter (Reference 2) is sufficient to support the NRC staff review of the CNP PRAs and therefore the NRC staff finds that it meets the requirements of 10 CFR 50.69(b)(2)(iii).

3.4.2.2 Peer Review of the PRA 3.4.2.2.1 Peer Review History of the Internal Events PRA In Section 3.3 of the LAR enclosure, the licensee states that the FPIE PRA model (which includes internal floods) was subjected to a full-scope peer review in July 2015, consistent with RG 1.200, Revision 2 (Reference 6) and process in NEI 05-04 (Reference 25). In the response to APLA-RAI-1 (Reference 2), the licensee confirmed that the 2015 version of the FPIE model served as the starting point for the current 2023 version of the FPIE model. As discussed in the LAR, the model underwent a focused-scope peer review in September 2017 and a subsequent closure review March 2018 for a PRA upgrade related to containment hydrogen analysis. Additional focused-scope peer reviews were performed in October 2016 and November 2020 focusing on the treatment of pre-initiator human reliability analysis (HRA) and the implementation of FLEX, respectively. An F&O closure review was performed in November 2021 using the NEI 05-04 process (Reference 25), the ASME/ANS PRA Standard (Reference 14), RG 1.200, Revision 2 (Reference 6) and Appendix X guidance (Reference 13).

The disposition of open finding-level F&Os for the FPIE PRA model are discussed in the next section of this safety evaluation. (There are no remaining open F&Os involving the internal flooding PRA model.)

The NRC staff finds that the FPIE PRA was peer reviewed in accordance with PRA Standard ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (the 2009 PRA standard) (Reference 14), as endorsed in RG 1.200, Revision 2, and that closed F&Os were assessed by an independent assessment team using the guidance in Appendix X.

Therefore, the NRC staff concludes that the CNP FPIE PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and is acceptable for risk-informed safety categorization of SSCs.

3.4.2.2.2 Disposition of Open Finding-Level F&Os for the Internal Events PRA In Attachment 3 of Enclosure 2 to the LAR, the licensee described the open F&O findings for the FPIE PRA and provided a disposition for each of them. In response to APLA-RAI-1 (Reference 2), the licensee identified and confirmed the PRA Standard supporting requirements (SRs) that are considered to be Met at Capability Category I (CC-I) only. As discussed in RG 1.200, Revision 2 (Reference 6), the staff considers meeting Capability Category II (CC-II) of the ASME/ANS PRA standard to be generally acceptable for risk-informed applications.

However, CC-I may be sufficient for some applications, depending on the technical requirement and the specific application.

In the supplement (Reference 2), the licensee provided further justification that conservatisms associated with the LERF analysis (i.e., those associated with SRs met at CC-I), including the containment hydrogen analysis, would not mask risk insights and therefore categorization of SSCs would not be adversely impacted. In response to APLA-RAI-1 (Reference 2), the licensee provided a detailed explanation of the identified F&Os, the associated SRs, and a basis for why categorization would not be significantly impacted. The licensee also explained why the issues identified in the F&Os would not constitute key assumptions or sources of uncertainty for the application. The NRC staff finds that the PRA modeling aspects which resulted in LERF SRs being met at CC-I are not overly conservative and are modeled sufficiently to support risk-informed categorization in accordance with 10 CFR 50.69.

In the RAI response (Reference 2), the licensee also confirmed which systems still require updated walkdowns. The licensee explained that updated walkdowns are needed to fully resolve one open F&O and that the licensees process for identifying system design/operating changes would have captured any changes and incorporated them into the PRA so deviations between the current model and the current plant would not be expected. The NRC staff finds that the PRA configuration control process is sufficient to resolve potential discrepancies between the PRA model of record and the as-built, as-operated plant, with respect to the F&Os.

See Section 3.4.4 of this safety evaluation for additional discussion of the licensees PRA change control and periodic review process.

The NRC staff reviewed the LAR (Reference 1) and supplemental information (Reference 2),

and found that, based on the above, the licensee provided sufficient justification to disposition the open finding-level F&Os for the 10 CFR 50.69 application. Therefore, the NRC staff finds that the CNP FPIE PRA is acceptable to support the risk-informed categorization process consistent with RG 1.201, Revision 1 (Reference 8).

3.4.2.2.3 Peer Review History of the Fire PRA In Section 3.2.2 of the LAR enclosure, the licensee states that the internal fire PRA model was developed consistent with NUREG/CR-6850 (Reference 17), RG 1.200 (Reference 6) and the 2009 PRA Standard (Reference 14). The NRC notes that the CNP is also approved to use a risk-informed, performance--based fire protection program in accordance with 10 CFR 50.48(c)

(Reference 26) that relies on the fire PRA.

In Section 3.3 of the LAR enclosure, the licensee states that the fire PRA was subjected to a full-scope peer review in July 2010, consistent with RG 1.200, Revision 2 (Reference 6), and the process in NEI 07-12 (Reference 15) (with exception of the Qualitative Screening and Quantitative Screening elements as screening tasks were not performed in the fire PRA).

Additional focused-scope peer reviews were performed in November 2015, July 2017, July 2022, and February 2023. An F&O closure review was conducted in June 2023. The disposition of open finding-level F&Os for the Fire PRA model is discussed in the next section of this safety evaluation.

The NRC staff finds that the fire PRA was peer reviewed, consistent with the 2009 PRA Standard (Reference 14) as endorsed in RG 1.200, Revision 2 (Reference 6), and that closed F&Os have been assessed using the guidance in Appendix X. Therefore, the NRC staff concludes that the CNP fire PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and is acceptable for risk-informed safety categorization of SSCs.

3.4.2.2.4 Disposition of Open Finding-Level F&Os for the Fire PRA In Attachment 3 of Enclosure 2 to the LAR, the licensee described the open finding-level F&Os for the fire PRA and provided a disposition for each of them. The open F&Os were associated with documentation issues and conservatisms in the LERF analysis for the FPIE PRA. The LERF analysis SRs for FPIE PRA were limited to CC-I; therefore, the associated SR in the fire PRA was also limited to CC-I. As discussed in Section 3.4.2.2.2 of this safety evaluation, the NRC found that the FPIE LERF analysis finding-level F&O were adequately dispositioned for the application. The NRC also finds that the minor documentation issues identified in the LAR enclosure would not impact SSC categorization.

The NRC staff reviewed the LAR (Reference 1), as supplemented (Reference 2), and found that it provided sufficient justification to disposition the open F&Os for the 10 CFR 50.69 application.

Therefore, the NRC staff finds that the CNP fire PRA is acceptable to support the categorization process using the process endorsed by the NRC staff in RG 1.201, Revision 1 (Reference 8).

3.4.2.2.5 Peer Review History of the Seismic PRA In Enclosure 2, Section 3 of the LAR (Reference 1), the licensee stated that the seismic PRA model received a full-scope peer review in November 2018 using the ASME/ANS RA Sb-2013, Code Case 1 (Reference 18), which the NRC staff accepted for use by letter dated March 12, 2018 (Reference 28) and later endorsed in RG 1.200, Revision 3 (Reference 7). In response to a RAI (Reference 2), the licensee confirmed that it adhered to both Code Case 1 and the NRC staffs clarifications and qualifications in the development of the seismic PRA.

CNP performed a peer review of the seismic PRA using the peer review process described in NEI 12-13 and accepted, with limited comments, by the NRC as an approach to performing peer reviews on March 7, 2018 (Reference 29). A subsequent independent assessment for closure of F&Os was performed in August 2019, which resulted in closure of 32 of 45 finding-level F&Os.

The disposition of open finding level F&Os for the seismic PRA model are discussed in the next section of this safety evaluation.

3.4.2.2.6 Disposition of Open Finding-Level F&Os for the Seismic PRA Open and partially resolved finding-level F&Os associated with the seismic PRA are identified and dispositioned for impacts to the seismic PRA results in Attachment 3 of Enclosure 2 of the LAR (Reference 1). In the response to the RAI (Reference 2), the licensee provided further justification that open and partially resolved finding-level F&Os associated with PRA would not impact SSC categorization under 10 CFR 50.69. The licensee stated that, since the global structural failure mode is directly modeled to core damage in the PRA analysis and dominates the risk assessment, minor adjustments to structural modeling and variations in fragilities would not affect SSC categorization under 10 CFR 50.69. The staff finds the licensees evaluation to be acceptable because the global structural failure is the dominant contributor to the plant-level seismic risk, and a qualitative review of high-risk fragilities was performed by the licensee which demonstrates further adjustments would not impact SSC categorization.

Other open seismic PRA F&Os were associated with documentation issues and conservatisms in the LERF analysis for the FPIE PRA. Since the SR for FPIE PRA was limited to CC-I, the associated SR in the seismic PRA was also limited to CC-I. As discussed in Section 3.4.2.2.2 of this safety evaluation, conservatisms identified in the LERF analysis are not expected to impact categorization results. Staff also concluded that the documentation issues identified would not impact categorization.

Based on its review and the above conclusions, the NRC staff finds that the seismic PRA was peer reviewed consistent with RG 1.200 (Reference 6), and all finding-level F&Os were closed or were dispositioned as not impacting SSC categorization under 10 CFR 50.69. The NRC staff also finds that the licensees evaluation of key assumptions and sources of uncertainty for the seismic PRA is consistent with RG 1.200 (Reference 6). Therefore, the licensees seismic PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and is acceptable for risk-informed safety categorization of SSCs.

3.4.2.3 Credit for Diverse and Flexible Coping Strategy (FLEX) Equipment The NRC staff has identified challenges to incorporating FLEX equipment and strategies into a PRA model to support risk-informed decision-making. The NRC staff assessment of industry guidance for crediting mitigation strategies in PRA is documented in a memorandum dated May 6, 2022, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments (Reference 19).

In Section 3.2.8, Modeling of FLEX, of the LAR enclosure, the licensee states that FLEX strategies are credited in the CNP internal events and fire PRA models and that no credit is taken for FLEX strategies within the seismic PRA model. The licensee explains in the LAR how FLEX was implemented in the applicable PRA models and states that a focused-scope peer review was performed to review the implementation. The licensee also states that HRA was included as an element of the peer review which confirmed HRA modeling was consistent with the PRA standard. In addition, the licensee states that a gap assessment was performed against the NRC memorandum, dated May 6, 2022, and FLEX human event probabilities were updated as a result of this review. The staff also reviewed the discussion of uncertainty in Section 3.2.7 and Attachment 6 of the LAR and confirmed that FLEX modeling was not identified as a key source of uncertainty for the application to adopt 10 CFR 50.69.

Based on the above discussion, the NRC concludes that the licensee assessed FLEX modeling consistent with the NRC memorandum (Reference 19) for modeling of FLEX and treatment of uncertainties. The NRC staff finds that the uncertainties associated with the PRA modeling of FLEX equipment and operator actions have been adequately addressed consistent with NEI 00-04 guidance, as endorsed by RG 1.201, Revision 1, to support the CNP 10 CFR 50.69 categorization process.

3.4.2.4 Assessment of Assumptions and Approximations 3.4.2.4.1 Identification of Key Assumptions and Sources of Uncertainty In Section 3.2.7 and in Attachment 6 of the LAR enclosure (Reference 1), the licensee confirms that evaluations of uncertainty meet the intent of NUREG-1855, Revision 1 (Reference 9), to identify, screen, and characterize those sources of model uncertainty and related assumptions in the base PRA that are relevant to this application. Substep E-1.4, Qualitative Screening of Sources of Model Uncertainty and Related Assumptions, of NUREG-1855, provides guidance on a qualitative screening process that involves identifying and validating whether consensus models were used in the PRA to evaluate identified model uncertainties. In Attachment 6 of the LAR, the licensee confirmed that, for the CNP uncertainty analysis, some uncertainties and assumptions were screened based on the use of a consensus method, specifically for the reactor coolant pump shutdown seals PRA modeling. The NRC staff finds that the assessment performed to identify the key assumptions/sources of uncertainty is consistent with the guidance provided in NUREG-1855.

3.4.2.4.2 Treatment of the Key Assumptions and Sources of Uncertainty NUREG-1855, Revision 1 (Reference 9) and NEI 00-04 (Reference 11) provide guidance regarding how to address PRA uncertainties and on how to treat uncertainties associated with PRA in risk-informed decision-making. The guidance fosters an understanding of the uncertainties associated with PRA and their impact on the results of the PRA and provides an approach to addressing these uncertainties in the context of the decision-making.

In Section 3.2.7 of the LAR enclosure, the licensee states that uncertainty evaluations associated with the risk categorization process are addressed using the prescribed sensitivity studies discussed in Section 5 of NEI 00-04. In Attachment 6, Disposition of Key Assumptions/Sources of Uncertainty, of Enclosure 2 to the LAR (Reference 1), the licensee conservatively identified the modeling of Westinghouse Generation Ill Reactor Coolant Pump Shutdown Seals as a key source of modeling uncertainty. It was also explained in the LAR that the PRA models use consensus5 models for seal failure and a peer review found the modeling acceptable. Based on this review, the licensee determined that no additional sensitivity analyses are required to address CNP PRA model assumptions or sources of uncertainty for the 10 CFR 50.69 categorization process, as discussed in Section 3.2.7 of the LAR enclosure. The NRC staff finds that the licensee plans to perform sensitivity studies consistent with NEI 00-04 guidance to address the identified key assumptions and sources of uncertainty and to address PRA uncertainties consistent with NUREG-1855 is acceptable.

In response to APLAI-RAI-3 (Reference 2), the licensee provided further discussion of the treatment of parametric uncertainty and the results of a sensitivity analysis using the parametric mean values for CDF and LERF for the internal events and fire PRA models. This evaluation confirmed that total CDF and LERF values would remain below the risk acceptance guidelines in RG 1.174 and demonstrated that the SSC categorizations would likely remain unchanged if the parametric mean values were utilized as opposed to the point estimate values.

The NRC staff recognizes that the licensee will perform routine PRA changes and updates to assure the PRA continually reflects the as-built, as-operated plant in addition to changes made 5 As defined in NUREG-1855, a consensus model is a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group.

to the PRA to support the context of the analysis being performed (i.e., sensitivities).

Paragraphs 50.69(e) and (f) of 10 CFR stipulate the process for feedback and adjustment to assure configuration control is maintained for these routine changes and updates to the PRA.

The NRC staff evaluation of this activity is documented in Section 3.4.4 of this safety evaluation.

3.4.2.5 PRA Importance Measures and Integrated Importance Measures The scope of modeled hazards for CNP includes the internal events PRA (including internal floods), fire PRA, and seismic PRA. 10 CFR 50.69(c)(1)(ii) requires that the SSC functional importance be determined using an integrated, systematic process. Section 5.6 of NEI 00-04 (Reference 11), discusses the need for an integrated computation using available importance measures. It further states that the integrated importance measure essentially weights the importance from each risk contributor (e.g., internal events, fire, seismic PRAs) by the fraction of the total core damage frequency [or large early release frequency] contributed by that contributor.

In response to APLC-RAI-1a (Reference 2), the licensee addressed the integration of importance measures across hazards for the 10 CFR 50.69 categorization process. The licensee stated that the process to categorize each system will be consistent with the guidance in Section 5.6 of NEI 00-04.

In the response to APLC-RAI-1b (Reference 2), the licensee explained the derivation of importance measures (e.g., Fussell-Vesely (FV) and risk achievement worth (RAW)) from the PRA models and the justification that the method used to generate the integrated importance measures is appropriate for use in the categorization process. The licensee stated that events in the PRA that represent each SSC are selected consistent with NEI 00-04, Section 5. The licensee stated that for the purposes of performing the active component risk categorizations, only metrics associated with the performance of the mitigation capability of the components are considered. Component failures due to the occurrence of the hazard are not included in the mapping of basic events to components.

In the response to APLC-RAI-1c (Reference 2), the licensee explained the derivation of importance measures (e.g., FV and RAW) from the seismic PRA and discussed how discretized basic events are combined to develop representative importance measures and compared to the importance measure thresholds in NEI 00-04 (Reference 11). For the discretized basic events associated with the seismic PRA, the licensee explained that the cutsets are quantified separately for each of the seismic initiator bins and these cutsets are then combined to calculate basic event importances that are compared to the criteria from NEI 00-04 to determine the categorization of a specific component.

In the response to APLC-RAI-1d (Reference 2), the licensee explained how basic events in the seismic PRA that differ from basic events in other PRA models will be considered when performing an integral assessment of component performance. The licensee stated that all basic events associated with a particular system will be mapped to a system component, such that any basic events that appear only in a fire PRA or only in a seismic PRA will be mapped to one of the system components and any basic event importances will be included in the categorization of the component in both initial categorization and in any integrated assessment.

The NRC staff finds that the licensees use and treatment of importance measures is consistent with the guidance in NEI 00-04 (Reference 11) as endorsed in RG 1.201, Revision 1 (Reference 8). Therefore, the NRC staff concludes that the licensees approach meets 10 CFR 50.69(c)(1)(ii).

3.4.2.6 PRA Acceptability Conclusions Pursuant to 10 CFR 50.69(c)(1)(i), the categorization process must consider results and insights from a plant-specific PRA. The use of the internal events PRA (including internal floods), fire PRA, and seismic PRA to support SSC categorization is endorsed by RG 1.201, Revision 1 (Reference 8). The PRAs must be acceptable to support the categorization process and must be subjected to a peer review process. The NRC staff evaluation of the peer reviews for CNP is discussed in Section 3.4.2.2, above.

The NRC staff finds the licensee provided sufficient information required for the NRC staff to conclude that the licensees PRA, comprising an internal events PRA, a fire PRA, and a seismic PRA, is acceptable to support the categorization of SSCs using the process endorsed by the NRC staff in RG 1.201, Revision 1 (Reference 8). The key assumptions for the PRAs have been identified, consistent with the guidance in RG 1.200, Revision 2 (Reference 6) and NUREG-1855 (Reference 9), and addressed appropriately for this application. Therefore, the NRC staff concludes that the licensees PRA meets the requirements set forth in 10 CFR 50.69(c)(1)(i) and (ii).

3.4.3 Evaluation of the Use of Non-PRA Methods in SSC Categorization The licensees categorization process uses the following non-PRA methods:

IPEEE screening analysis for high winds, external floods, and other hazards (except seismic), updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 (Reference 14)

Safe Shutdown Risk Management program consistent with NUMARC 91-06 (Reference 12)

Passive Components: ANO-2 passive categorization (Reference 21)

The NRC staffs review of these methods is discussed below.

3.4.3.1 Other External Hazards Other external hazards include all non-seismic external hazards such as high winds, external floods, transportation, nearby facility accidents, and other hazards. In Section 3.2.4 of to the LAR (Reference 1), the licensee stated that in 2022 it performed a systematic evaluation of the external hazards using the guidance in ASME/ANS PRA Standard RA-Sa-2009 (Reference 14). Attachment 4 of the LAR concluded that the external hazards -

other than seismic - can be screened out; therefore, there was no need for further detailed PRA of these external hazards.

For external flooding, all flooding events except for a local intense precipitation (LIP) event were determined as part of the NRCs post-Fukushima 50.54(f) Request for Information (Reference 27) to be bounded by the current licensing basis. Because these events are bounded, they were able to be screened out using the initial preliminary screening criteria C1, Event damage potential is less than events for which the plant is designed as shown in LAR of Enclosure 2. For a LIP event, all components were identified whose failure to be in their normal position (e.g., doors in the closed position) or function appropriately (e.g., roof scuppers shedding water to not allow water to pool) to limit the ingress of water to the auxiliary and turbine buildings. The licensee has stated that all these identified components will be categorized as HSS to screen out the hazard from further consideration for 10 CFR 50.69 categorization.

For the screening of the external hazard of ice cover, the de-icing valves (1-WMO-16 and 2-WMO-26) are categorized as HSS. In response to APLC-RAI-4a (Reference 2), the licensee described how the de-icing system is operated and how the de-icing valves are used when cold weather conditions lead to the buildup of ice on Lake Michigan. The licensee concluded, in LAR of Enclosure 2 as supplemented, that criteria C4 and C5 are applicable to dispositioning the hazard for 50.69.

For the external hazard of release of chemicals from on-site storage, the NRC issued APLC-RAI-4b to determine if a hydrazine spill had been considered outside of storage. In response to APLC-RAI-4b (Reference 2), the licensee described how the current control room habitability evaluation for a postulated hydrazine release bounds the potential impacts of an in-transit hydrazine spill. The licensee also concluded that further consideration of these external hazard failure mechanisms would not impact SSC classification under 10 CFR 50.69.

For the external hazard of turbine-generated missiles the NRC issued APLC-RAI-4c to determine if a failure of the high-pressure turbines (not just low-pressure turbines) had been considered. In response to APLC-RAI-4c (Reference 2), the licensee described how the analysis of the low-pressure turbine missiles bounds the potential impacts of high-pressure turbine missiles. The licensee also concluded that further consideration of these external hazard failure mechanisms would not impact SSC classification under 10 CFR 50.69.

The NRC staff reviewed the licensees progressive screening approach for external hazards described in Enclosure 2 to the LAR (Reference 1) and the additional information provided in response to the staffs requests for additional information. The NRC staff notes that the progressive screening criteria used in Attachment 4 of Enclosure 2 to the LAR are the same criteria presented in supporting requirements for screening external hazards of the NRC-endorsed 2009 PRA Standard (Reference 14). Based on its review, the NRC staff concludes that the licensees treatment of other external hazards is acceptable and meets 10 CFR 50.69(c)(1)(ii).

3.4.3.2 Shutdown Consistent with the guidance in NEI 00-04 (Reference 11), the licensee proposed using a shutdown safety assessment based on NUMARC 91-06 (Reference 12). This industry guidance addresses considerations for maintaining defense-in-depth for the five key safety functions during shutdown, namely, decay heat removal capability, inventory control, power availability, reactivity control, and containment. The guidance also specifies that a defense-in-depth approach should be used with respect to each defined shutdown key safety function. This is accomplished by designating a running and an alternative system/train to accomplish the given key safety function.

The use of NUMARC 91-06 described by the licensee in the submittal is consistent with the guidance in NEI 00-04 as endorsed by the NRC in RG 1.201, Revision 1 (Reference 8). The approach uses an integrated and systematic process to identify HSS components. Therefore, the NRC staff finds that the licensees use of NUMARC 91-06 is acceptable, and meets the requirements set forth in 10 CFR 50.69(c)(1)(ii).

3.4.3.3 Component Safety Significance Assessment for Passive Components Passive components are not modeled in the PRA; therefore, a different assessment method is necessary to assess the safety significance of these components. Passive components are those components having only a pressure retaining function. This process also addresses the passive function of active components such as the pressure/liquid retention of the body of a motor-operated valve.

In Section 3.1.2, Passive Categorization Process, of the LAR enclosure (Reference 1), the licensee proposed using a categorization method for passive components which was not cited in NEI 00-04 (Reference 11) or RG 1.201, Revision 1 (Reference 8). This method was approved by the NRC for ANO-2 (Reference 21).

The ANO-2 methodology is a risk-informed safety classification and treatment program for repair/replacement activities for Class 2 and 3 pressure retaining items and their associated supports (exclusive of Class CC and MC items) using a modification of ASME Code Case N-660 (Reference 22).

In Section 3.1.2 of the LAR enclosure (Reference 1) the licensee states, in part, that the use of the ANO-2 risk-informed repair/replacement activities (RI-RRA) method was previously approved to be used for a 10 CFR 50.69 application by NRC in the final Safety Evaluation for Vogtle dated December 17, 2014 (Reference 31). The RI-RRA method as approved for use at Vogtle for 10 CFR 50.69 does not have any plant specific aspects and is generic. The ANO-2 RI-RRA methodology relies on the conditional core damage and large early release probabilities associated with pipe ruptures. Safety significance is generally measured by the frequency and the consequence of a particular event, in this case, a pipe rupture. However, treatment requirements (including repair and replacement) only affect the frequency of passive component failure. Categorizing the significance of pipe failure based on the consequence of failure alone is conservative compared to the more general measure (which includes rupture frequency). The categorization will not be affected by changes in frequency arising from changes to the treatment.

In Section 3.1.2 of the LAR enclosure (Reference 1), the licensee states, in part, that:

The passive categorization process is intended to apply the same risk-informed process accepted by the NRC in ANO-2-R&R-004 for the passive categorization of Class 2, 3, and non-class components [Reference 21]. All ASME Code Class 1 SSCs with a pressure retaining function, as well as categorized supports, will be assigned HSS for passive categorization which will result in HSS for its risk-informed safety classification and cannot be changed by the IDP.

The proposed categorization methodology and ANO-2 precedent repair/replacement methodology do not allow an ASME Code Class 1 pressure retaining SSCs (and associated supports, etc.) to be recategorized from HSS to LSS.

In Section 3.1.2 of the LAR enclosure (Reference 1), as supplemented by the response to APLA-RAI-2 (Reference 2), the licensee also provided the following clarifications for supports:

Consistent with the ANO-2 Safety Evaluation pipe supports were not required to be in the scope of the evaluation but may be included in the scope at the licensee's discretion. Component supports, if categorized, are assigned safety based upon one of the following approaches:

Supports should have the same safety significance as the highest passively ranked component within the bounds of the associated analytical pipe stress model. Consistent with NEI 00-04, an HSS determination by the passive categorization process cannot be changed by the IDP.

A combination of restraints or supports such that the LSS piping and associated SSCs attached to the HSS piping are included in scope up to a boundary point that encompasses at least two supports in each of three orthogonal directions.

The RAI response explains that the latter bullet is based on the guidance in the standard review plans for license renewal and subsequent license renewal (References 23 and 24).

The NRC staff finds that the use of the repair/replacement methodology is consistent with the methodology evaluated and accepted previously in the case of ANO-2, as documented in the safety evaluation for acceptable and appropriate for passive component categorization of Class 2 and Class 3 SSCs (Reference 21). The staff also finds that the clarifications regarding the treatments of supports are acceptable and consistent with current guidance. Accordingly, the NRC staff finds the licensees proposed approach for passive categorization is acceptable for the 10 CFR 50.69 SSC categorization process.

3.4.3.4 Assembly of Plant-Specific Inputs The licensees risk categorization process uses PRAs to assess risks from internal events (including internal floods), internal fire, and seismic hazards. Non-PRA methods are used to assess shutdown safety, passive component risk, and the significance of external hazards. For those that depart from the methodology prescribed in NEI 00-04, additional NRC staff review is discussed in Section 3.4.3.3 of this safety evaluation.

The process used by the licensee is described in the LAR (Reference 1), as supplemented (Reference 2). The process for collecting and organizing information at the system level for defining boundaries, functions, and components is consistent with NEI 00-04 (Reference 11).

Because the process is consistent with NEI 00-04 as clarified and endorsed in RG 1.201, Revision 1 (Reference 8), the NRC staff finds that the process meets the requirements set forth in 10 CFR 50.69(c)(1)(v).

3.4.3.5 Risk Sensitivity Study (NEI 00-04, Section 8)

In Section 3.1.1 of the LAR enclosure (Reference 1), the licensee acknowledges that RG 1.201, Revision 1 (Reference 8) states, the implementation of all processes described in NEI 00-04 (i.e., Sections 2 through 12) is integral to providing reasonable confidence and all aspects of NEI 00-04 must be followed to achieve reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(iv). The LAR goes on to state that an unreliability factor of three will be used for the sensitivity studies described in Section 8 of NEI 00-04.

In Section 3.2.7, PRA Uncertainty Evaluations, of the LAR enclosure, the licensee further confirms that a cumulative sensitivity study will be performed where the failure probabilities (unreliability and unavailability, as appropriate) of all LSS components modeled in all identified PRA models for all systems that have been categorized are increased by a factor of three. In Section 3.4 of the LAR enclosure, the licensee states, in part:

Sensitivity studies described in NEI 00-04 Section 8 will be used to confirm that the categorization process results in acceptably small increases to core damage frequency (CDF) and large early release frequency (LERF).

Subsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

The cumulative sensitivity study, together with the periodic review process, discussed in Section 3.4.4.1 of this safety evaluation, assures that the potential cumulative risk increase from the categorization is maintained acceptably low. Section 3.2.7 of the LAR enclosure further states, The performance monitoring process monitors the component performance to ensure that potential increases in failure rates of categorized components are detected and addressed, through CNP's Corrective Action Program, before reaching the rate assumed in the sensitivity study.

The NRC staff finds that the licensee will perform the risk sensitivity study consistent with the guidance in Section 8 of NEI 00-04 (Reference 11) and, therefore, will assure that the potential cumulative risk increase from the categorization is maintained acceptably low, as required by 10 CFR 50.69(c)(1)(iv).

3.4.3.6 Integrated Decision-Making Appendix B of SRP Section 19.2 (Reference 10) provides guidance and the NRC staff expectations for the licensees integrated decision-making process. The appendix states in part, Risk-informed applications are expected to require a process to integrate traditional engineering and probabilistic considerations to form the basis for acceptance. NEI 00-04 guidance (Reference 11) identifies two steps in the categorization process: (1) Preliminary Engineering Categorization of Function, and (2) IDP Review and Approval. These call for the integrated assessment of the traditional engineering analyses and the risk results. Risk results come from the PRA and non-PRA assessments. Integrated decision-making is performed to make a determination of the safety significance of SSCs for their categorization. The staff reviews the two steps to ensure the processes is well-defined, systematic, repeatable, and scrutable.

3.4.3.7 Preliminary Engineering Categorization of Function (NEI 00-04, Section 7)

In Section 3.1.1 of the LAR enclosure (Reference 1), the licensee acknowledged the NRC staffs clarification of NEI 00-04 that if any SSC is identified as HSS from either the integrated PRA component safety significance assessment (Section 5 of NEI 00-04) or the defense-in-depth assessment (Section 6), the associated system function(s) would be identified as HSS. The licensee also states, Once a system function is identified as HSS, then all the components that support that function are preliminary HSS. The IDP must intervene to assign any of these HSS function components to Low Safety Significant (LSS).

Based on the above, the NRC staff finds that the above description provided by the licensee for the preliminary categorization of functions is consistent with NEI 00-04 (Reference 11) and is therefore acceptable.

3.4.3.8 IDP Review and Approval (NEI 00-04, Sections 9 and 10)

In Section 3.1.1 of the LAR enclosure (Reference 1), the licensee states:

The Integrated Decision-Making Panel (IDP) will be composed of a group of at least five experts who collectively have expertise in plant operation, design engineering, system engineering, safety analysis, and probabilistic risk assessment. At least three members of the IDP will have a minimum of five years of experience at the plant, and there will be at least one member of the IDP who has a minimum of three years of experience in the modeling and updating of the plant-specific PRA.

The IDP will be trained in the specific technical aspects and requirements related to the categorization process. Training will address at a minimum the purpose of the categorization; present treatment requirements for SSCs including requirements for design basis events; PRA fundamentals; details of the plant-specific PRA including the modeling, scope, and assumptions, the interpretation of risk importance measures, and the role of sensitivity studies and the change-in-risk evaluations; and the defense-in-depth philosophy and requirements to maintain this philosophy.

The guidance in NEI 00-04 (Reference 11), as endorsed by the NRC in RG 1.201 (Reference 8), provides confidence that the IDP expertise is sufficient to perform the categorization and that the results of the different evaluations (PRA and non-PRA) are used in an integrated, systematic process as required by 10 CFR 50.69(c)(1)(ii). Based on the above, the NRC staff finds that the licensee's proposed IDP would have the expertise to meet the requirements in 10 CFR 50.69(c)(2) and that the additional descriptions of the IDP characteristics, training, processes, and decision guidelines are consistent with NEI 00-04, as endorsed by the NRC in RG 1.201.

3.4.3.9 Conclusion for Key Principle 4 The NRC staff reviewed the acceptability of the licensees internal events PRA (including internal floods), fire PRA, and seismic PRA. The staff also reviewed the use of PRA importance measures and integrated importance measures, the use of non-PRA methods, risk sensitivity studies, and integrated decision-making. Based on these reviews and the findings described in Sections 3.4.2 and 3.4.3 above, the NRC staff has determined that the proposed change satisfies the fourth key principle for risk-informed decision-making identified in RG 1.174, Revision 3 (Reference 4) and ensures that any potential increases in risk are small as required by 10 CFR 50.69(c)(1)(iv).

3.4.4 Key Principle 5: Monitor the Impact of the Proposed Change NEI 00-04 (Reference 11) provides guidance that includes programmatic configuration control and a periodic review to ensure that the all aspects of the 10 CFR 50.69 program (including traditional engineering analyses) and PRA models used to perform the risk assessment continue to reflect the as-built-as-operated plant and that plant modifications and updates to the PRA are continually incorporated.

3.4.4.1 Periodic Review Paragraph 50.69(e) of 10 CFR requires that periodic updates to the licensees PRA and SSC categorization must be performed. Changes over time to the PRA and to the SSC reliabilities are inevitable, and such changes are recognized by the 10 CFR 50.69(e) requirement for periodic updates.

In Section 3.2.6, PRA Maintenance and Updates, of the LAR enclosure (Reference 1), the licensee described the process for maintaining and updating the CNP PRA models used for the 10 CFR 50.69 categorization process. Consistent with NEI 00-04, the licensee confirmed that the CNP risk management process ensures that the PRA models used in this application continue to reflect the as-built and as-operated plant. The licensees process includes provisions for: monitoring issues affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience); assessing the risk impact of unincorporated changes; and controlling the model and associated computer files.

The process also includes reevaluating previously categorized systems to ensure the continued validity of the categorization. Routine PRA updates are performed every two refueling outages at a minimum.

Section 12.1 of NEI 00-04, Revision 0, states, in part, [s]cheduled periodic reviews (e.g. once per two fuel cycles in a unit) should evaluate new insight resulting from available risk information. In Section 3.5, Feedback and Adjustment Process, of the LAR enclosure (Reference 1), the licensee states, in part:

Scheduled periodic reviews will be completed at least once every two refueling cycles and will evaluate new insights resulting from available risk information changes (i.e., PRA model or other analysis used in the categorization), design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated.

The NRC staff finds the risk management process described by the licensee in the LAR is consistent with the guidance in Section 12 of NEI 00-04 (Reference 11). Considering the above, the staff has determined that the proposed change satisfies the fifth key principle for risk-informed decision-making identified in RG 1.174 (Reference 4).

3.4.4.2 Program Documentation and Change Control Paragraph 50.69(f) of 10 CFR requires program documentation, change control, and records. In Section 3.2.6, PRA Maintenance and Updates, of the LAR enclosure (Reference 1), the licensee stated that it will implement a process that addresses the requirements in Section 11 of NEI 00-04 (Reference 11), pertaining to program documentation and change control records. In Section 3.1.1 of the LAR enclosure, the licensee states that the RISC categorization process documentation will include the following 10 elements:

1.

Program procedures used in the categorization

2.

System functions, identified and categorized with the associated bases

3.

Mapping of components to support function(s)

4.

PRA model results, including sensitivity studies

5.

Hazards analyses, as applicable

6.

Passive categorization results and bases

7.

Categorization results including all associated bases and RISC classifications

8.

Component critical attributes for HSS SSCs

9.

Results of periodic reviews and SSC performance evaluations

10. IDP meeting minutes and qualification/training records for the IDP members The NRC staff also recognizes that for facilities licensed under 10 CFR Part 50, Appendix B Criterion VI, Document Control, procedures are considered formal plant documents requiring that [m]easures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality.

The elements provided in Section 3.1.1 of the LAR enclosure, in addition to the list of items provided in Attachment 1 of Enclosure 2 to the LAR, will ensure that the CNP 10 CFR 50.69 categorization process will be documented in formal licensee procedures. This is consistent with Section 11 of NEI 00-04, as endorsed by the NRC in RG 1.201, Revision 1 (Reference 8).

Based on the above, the NRC staff finds that the procedures will be sufficient for meeting the 10 CFR 50.69(f) requirement for program documentation, change control, and records.

4.0 CHANGES TO THE OPERATING LICENSE Based on the NRC staffs review of the LAR, as supplemented, the staff identified specific actions, as described below, that are necessary to support the NRC staffs conclusion that the proposed program meets the requirements in 10 CFR 50.69 and the guidance in NEI 00-04 (Reference 11) as endorsed in RG 1.201, Revision 1 (Reference 8). Note: Additional actions (e.g., final procedures and proposed alternative treatment) need not, and have not been submitted, or reviewed by the NRC staff for issuance of the safety evaluation but will be completed before implementation of the program as specified in the 10 CFR 50.69 rule.

The NRC staffs finding on the acceptability of the PRA evaluation in the licensees proposed 10 CFR 50.69 process is conditioned upon the license condition provided below.

The licensee proposed to add the following license condition to CNP RFOL Nos. DPR-58 and DPR-74:

The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC) RISC-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic hazards; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2, Class 3, and non-class SSCs and their associated supports; the results of non-PRA evaluations that are based on the Individual Plant Evaluation-External Events (IPEEE) Screening Assessment for External Hazards updated using the external hazard screening significance process identified in the ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; as specified in License Amendment No. [XXX] dated [DATE].

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

The NRC staff finds that the proposed license condition is acceptable because it adequately implements 10 CFR 50.69 using models, methods, and approaches consistent with the applicable NRC and NRC-endorsed guidance.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the State of Michigan official was notified of the proposed issuance of the amendments on March 12, 2025. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, as published in Federal Register on May 14, 2024 (89 FR 41999), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1.

Ferneau, K.J., Indiana Michigan Power Company, letter to the U.S. Nuclear Regulatory Commission, Donald C. Cook Nuclear Plant Units 1 and 2 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated March 6, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24073A234).

2.

Dailey, S. A., Indiana Michigan Power Company, letter to the U.S. Nuclear Regulatory Commission, Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Response to Request for Additional Information (RAI) for License Amendment Request (LAR) for Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, dated February 6, 2025 (ML25037A168).

3.

Wall, S. P., U.S. Nuclear Regulatory Commission, email to M. K. Scarpello, Indiana Michigan Power Company, Final RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Adoption of 10 CFR 50.69 (EPID No. L-2024-LLA-0025), dated December 30, 2024 (ML24366A003).

4.

U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018 (ML17317A256).

5.

U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 2, May 2011 (ML100910006).

6.

U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2, March 2009 (ML090410014).

7.

U.S. Nuclear Regulatory Commission, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 3, December 2020 (ML20238B871).

8.

U.S. Nuclear Regulatory Commission, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, Regulatory Guide 1.201, Revision 1, May 2006 (ML061090627).

9.

U.S. Nuclear Regulatory Commission, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, NUREG-1855, Revision 1, March 2017 (ML17062A466).

10.

U.S. Nuclear Regulatory Commission, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, NUREG-0800, Chapter 19, Section 19.2, June 2007 (ML071700658).

11.

Nuclear Energy Institute, 10 CFR 50.69 SSC Categorization Guideline, NEI 00-04, Revision 0, July 2005 (ML052910035).

12.

Nuclear Management and Resources Council, Inc., Guidelines for Industry Actions to Assess Shutdown Management, NUMARC 91-06, December 1991 (ML14365A203).

13.

Anderson, V. K., Nuclear Energy Institute, letter to S. Rosenberg, U.S. Nuclear Regulatory Commission, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os), dated February 21, 2017 (ML17086A431).

14.

American Society of Mechnical Engineers / American Nuclear Society, Addenda to ASME/ANS RA-S-2008, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, New York, NY, February 2009.

15.

Nuclear Energy Institute, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, NEI 07-12, Revision 0, Draft H, November 2008 (ML083430464).

16.

Electric Power Research Institute, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, EPRI 1026511, December 2012.

17.

U.S. Nuclear Regulatory Commission and Electric Power Research Institute, NUREG/CR-6850EPRI 1011989, Fire PRA Methodology for Nuclear Power Facilities, Volumes 1, Summary & Overview, and 2, Detailed Methodology, September 30, 2005 (ML15167A401 and ML15167A411, respectively).

18.

American Society of Mechanical Engineers / American Nuclear Society, ASME/ANS RA-S Case 1, Case for ASME/ANS RA-Sb-2013 Standard for Level 1 /

Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Applications, ASME, New York, NY, November 22, 2017.

19.

Zoulis, A. M., U.S. Nuclear Regulatory Commission, memorandum to M. Franovich, U.S. Nuclear Regulatory Commission, Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments, dated May 6, 2022 (ML22014A084).

20.

Electric Power Research Institute, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, EPRI 1016737, December 2008.

21.

Markley, M. T., U.S. Nuclear Regulatory Commission, letter to Vice President, Operation, Entergy Operations, Inc., Arkansas Nuclear One, Unit 2 - Approval of Request for Alternative ANO-2 R&R-004, Revision 1, Request to Use Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 and 3 Moderate and High Energy Systems (TAC No. MD5250), dated April 22, 2009 (ML090930246).

22.

American Society of Mechanical Engineers, Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities, ASME Code Case, N-660, July 2002.

23.

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24.

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25.

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26.

Wengert, T. J., U.S. Nuclear Regulatory Commission, letter to L.J. Weber, Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in accordance with 10 CFR 50.48(c) (TAC Nos. ME6629 and ME6630), dated October 24, 2013 (ML13140A398).

27.

Gebbie, J. P., Indiana Michigan Power Company, letter to the U.S. Nuclear Regulatory Commission, Donald C. Cook Nuclear Plant Units 1 and 2 Response to March 12, 2012, Request for Information, Enclosure 2, Recommendation 2.1: Flooding, Required Response 2, Hazard Reevaluation Report, dated March 6, 2015 (ML15069A334).

28.

Thomas, B. E., U.S. Nuclear Regulatory Commission, letter to C. R. Grantom and R. J.

Budnitz, U.S. Nuclear Regulatory Commission Acceptance of ASME/ANS RA-S Case 1, dated March 12, 2018 (ML18017A964).

29.

Franovich, M., U.S. Nuclear Regulatory Commission, letter to G. Krueger, U.S. Nuclear Regulatory Commission Acceptance of Nuclear Energy Institute (NEI) Guidance NEI 12-13, External Hazards PRA Peer Review Process Guidelines (August 2012),

dated March 7, 2018 (ML18025C025).

30.

Federal Register notice (69 FRN 68008, 68028-68029), dated November 22, 2004, related to Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors

31.

Martin, R., U.S. Nuclear Regulatory Commission, letter to Regulatory Affairs Director, Southern Nuclear Operating Company, Inc., Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Re: Use of 10 CFR 50.69 (TAC Nos. ME9472 and ME94473), dated December 17, 2014 (ML14237A034).

Principal Contributors: David Gennardo, NRR Mike Swim, NRR Adrienne Brown, NRR Sunwoo Park, NRR Edmund Kleeh, NRR Cory Parker, NRR Ming Li, NRR Gurjendra Bedi, NRR Stephen Cumblidge, NRR Brian Lee, NRR Fred Forsaty, NRR Date: April 9, 2025

ML25071A389

• via memo OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DRA/APLA/BC*

NAME SWall SRohrer RPascarelli DATE 03/12/2025 03/13/2025 03/04/2025 OFFICE NRR/DRA/APLC/BC*

NRR/DEX/EEEB/BC*

NRR/DEX/EICB/BC*

NAME ANeuhausen WMorton FSacko DATE 03/04/2025 03/04/2025 03/04/2025 OFFICE NRR/DEX/EMIB/BC*

NRR/DSS/SNSB/BC*

NRR/DSS/SCPB/BC*

NAME SBailey DMurdock MValentin DATE 03/04/2025 03/04/2025 03/04/2025 OFFICE NRR/DNRL/NVIB/BC*

NRR/DNRL/NPHP/BC*

OGC - NLO NAME ABuford MMitchell DRoth DATE 03/04/2025 03/04/2025 04/03/2025 OFFICE NRR/DORL/LPL3/BC(A)

NRR/DORL/LPL3/PM NAME IBerrios SWall DATE 04/08/2025 04/09/2025